The disposal of large quantities of toxic materials such as high level radioactive wastes stored in spent reactor storage pools, or generated in the reprocessing of spent nuclear power reactor fuel, or generated in the operation and maintenance of nuclear power plants, is a problem of considerable importance to the utilization of nuclear power. It is generally accepted that the most promising approach is to convert these radioactive wastes to a dry solid form which would render such wastes chemically, thermally and radiolytically stable.
The problem of dry solid stability of radioactive wastes is closely related to the safety of human life on earth for a period of more than 20,000 years. For example, radioactive wastes usually contain the isotopes Sr.sup.90, Pu.sup.239, and Cs.sup.137 whose half lives are 28 years, 24,000 years, and 30 years, respectively. These isotopes alone pose a significant threat to life and must be put into a dry, solid form which is stable for thousands of years. The solid radioactive waste form must be able to keep the radioactive isotopes immobilized for this length of time, preferably even in the presence of an aqueous environment. The radioactive wastes are produced in high volumes and contain long-lived, intermediate-lived, and short-lived radioactive ions and some non-radioactive ions. These solutions can be highly corrosive and it is difficult, if not impractical, to reduce them to concentrated forms for further processing or storage.
The two most popular types of commercial reactors both of which produce low level wastes are the Boiling Water Reactor (B.W.R.) and the Pressurized Water Reactor (P.W.R.). In a typical Pressurized Water Reactor (P.W.R.), pressurized light water circulates through the reactor core (heat source) to an external heat sink (steam generator). In the steam generator, where primary and secondary fluids are separated by impervious surfaces to prevent contamination, heat is transferred from the pressurized primary coolant to secondary coolant water to form steam for driving turbines to generate electricity. In a typical Boiling Water Reactor (B.W.R.), light water circulates through the reactor core (heat source) where it boils to form steam that passes to an external heat sink (turbine and condenser). In both reactor types, the primary coolant from the heat sink is purified and recycled to the heat source.
The primary coolant and dissolved impurities are activated by neutron interactions. Materials enter the primary coolant through corrosion of the fuel elements, reactor vessel, piping, and equipment. Activation of these corrosion products adds radioactive nuclides to the primary coolant. Corrosion inhibitors, such as lithium, are added to the reactor water. A chemical shim, boron, is added to the primary coolant of most P.W.R.'s for reactivity control. These chemicals are activated and add radionuclides to the primary coolant. Fission products diffuse or leak from fuel elements and add nuclides to the primary coolant. Radioactive materials from all these sources are transported around the system and appear in other parts of the plant through leaks and vents as well as in the effluent streams from processes used to treat the primary coolant. Gaseous and liquid radioactive wastes (radwaste) are processed within the plant to reduce the radioactive nuclides that will be released to the atmosphere and to bodies of water under controlled and monitored conditions in accordance with federal regulations.
The principal methods or unit operations used in the treatment of liquid radwaste at nuclear power plants are filtration, ion exchange, and evaporation.
Liquid radwastes in a P.W.R. are generally segregated into five categories according to their physical and chemical properties as follows:
a. Clean Waste includes liquids which are primarily controlled releases and leaks from the primary coolant loop and associated equipment. These are liquids of low solids content which are treated in the reactor coolant treatment system. PA1 b. Dirty or Miscellaneous Waste includes liquids which are collected from the containment building, auxiliary building, and chemical laboratory; regeneration solutions from ion-exchange beds; and solutions of high electrical conductivity and high solids content from miscellaneous sources. PA1 c. Steam Generator Blowdown Waste is condensate from the steam that is removed (blowdown) periodically to prevent excessive solids buildup. PA1 d. Turbine Building Drain Waste is leakage from the secondary system that is collected in the turbine building floor sump. PA1 e. Detergent Waste includes liquids from the laundry, personnel decontamination showers, and equipment decontamination. PA1 a. High-Purity Waste includes liquids of low electrical conductivity (&lt;50 .mu.mho/cm) and low solids content, i.e., reactor coolant water that has leaked from the primary reactor system equipment, the drywell floor drain, condensate demineralizer backwash, and other sources of high-quality water. PA1 b. Low-Purity Waste includes liquids of electrical conductivity in excess of 50 .mu.mho/cm and generally less than 100 .mu.mho/cm; i.e., primarily water from floor drains. PA1 c. Chemical Waste includes solutions of caustic and sulfuric acid which are used to regenerate ion exchange resins as well as solutions from laboratory drains and equipment decontamination. PA1 d. Detergent Waste includes liquids from the laundry and personnel decontamination showers.
Liquid radwastes in a B.W.R. are generally segregated into four categories according to their physical and chemical properties as follows:
The liquid radwastes from both types of reactors are highly dilute solutions of radioactive cations, and other dissolved radioactive materials as well as undissolved radioactive particles or finely divided solids.
A practical process for disposing of radioactive materials in a dry solids form having high resistance to leaching and other forms of chemical attack would not only be suitable for the disposal of radioactive nuclear wastes, but also for the fabrication of radioactive sources useful in industry, medicine, and in the laboratory.
Heretofore, there did not exist any practical, foolproof means for the safe disposal, storage and immobilization of pernicious radioactive waste material. Present day storage containers do not provide sufficient isolation and immobilization of such radioactive material, sufficient long-term resistance to chemical attack by the surroundings, and sufficient stability at high temperature.
Currently low level radioactive waste, that is radioactive waste generated at reactor sites, is disposed of in the following manner:
(A) The dead ion exchange resin containing radioactive waste is mixed with cement and cast in forty gallon barrels.
(B) The bottoms from evaporators which contain the radioactive contaminated boric acid and the solutions used to regenerate the ion exchange columns are mixed with cement powder and cast in forty gallon barrels.
(C) The filters containing particulate forms of radioactive waste are usually encased in cement in barrels.
These cement barrels are transported to low level radioactive waste sites and buried six feet deep in the ground. At least one of the sites is in the United States Eastern States and exposed to substantial rainfall. In Europe, these barrels are buried at sea. In both cases water will first corrode the metal then the cement and will relatively quickly expose the radioactive ions for leaching into the ground water or sea water. Because the U.S. burials are only a few feet deep, the contaminated water can readily intermix with streams, lakes and rivers, thus, entering the ecosphere. The rationale for this practice is the assumption that upon sufficient dilution the radioactivity becomes harmless.
Some of the most serious nuclear wastes are cesium and strontium which are biologically similar to sodium and calcium. They have thirty year half lives indicating that they should be isolated from the exosphere for at least three hundred years (ten half lives). At Bikini, the experts assumed that dilution had made the island inhabitable after decades in which no atomic explosions were performed, yet when the population was returned to the island its health was deleteriously effected. It has since been realized that plants and animal life biologically reconcentrate these radioactive elements back up to dangerous levels.
Thus, the "safe" concentration of radioactive waste must be much lower than accepted values and a more durable substitute for cement is needed. In one aspect, the present invention presents a safe alternative to the cement-solidification of low level waste.
Another route heretofore suggested is the so-called dry solids approach which involves the fixation of the waste materials in glasses via mixing with glass-forming compositions and melting to form glasses. This approach offers some improvement regarding isolation and decrease in the rate of release of radioactive elements when the outer envelopes or containers are destroyed. Further, such glasses remain relatively more stable at high temperatures than plastic and are generally more chemically durable in saline solutions than are metals. Glasses with high chemical durability and low alkali ion conductivity suitable for this prior art technique are formed at very high temperatures, e.g., 1800.degree. C. and higher. Prior processes utilizing such high melting glass-forming compositions are economically unsound and moreover, cause a dangerous problem due to the risk of volatilization of pernicious radioactive materials. Furthermore, this procedure is restricted to dry solid radioactive wastes and provides no solution to the high volumes of liquid radioactive wastes produced by the operations and maintenance of nuclear reactors, by the current practice of storing spent fuels in pools of water, and by spent reactor fuel recovery systems.
In view of the overall difficulties of handling radioactive material, and especially in view of the danger of volatilization of radioactive material into the atmosphere, attention has been directed to using glass compositions having relatively low melting temperatures, that is to say, using glass compositions with SiO.sub.2 contents as low as 27 weight percent. While the problem of volatilization of radioactive materials is reduced, it is not completely controlled. Moreover, the resultant glass composition exhibits greatly reduced chemical durability and increased ion diffusion rates for the radioactive materials present therein. The greater this diffusion rate, the lower is the ability of the glass to keep the radioactive materials immobilized in its matrix. For long-term containment of radioactive waste, demanded under present day standard, these prior glass compositions are inadequate.
U.S. Pat. No. 3,640,888 teaches the production of neutron sources by encapsulating californium-252 in glass using the steps of packing an open-ended vitreous tube a porous powder of quartz having an organic liquid ion exchange material sorbed thereon, passing an aqueous solution containing californium-252 cation through the powdered quartz, drying and heating the powdered quartz and tube in air to oxidize and volatilize the organic liquid ion exchange material resulting in the non-volatile oxide of californium-252, and then fusing the tip and powder contents to form a vitreous body containing the californium-252 oxide. The patent, however, does not disclose, teach or suggest the use of porous glass containing silicon-bonded cation-oxy groups which are exchangeable by radioactive cations in aqueous solution nor does it disclose or suggest any method or technique for concentrating and safely disposing of radioactive wastes.
U.S. Pat. No. 1,533,794 teaches the packaging of radium emanations in a glass capillary tube followed by sealing the ends of the tube and thus enclose emanations previously introduced into the tube. There is no teaching, however, of any method for concentrating and/or immobilizing radwaste.
U.S. Pat. Nos. 2,336,227; 2,340,013; 2,522,524; 3,364,148 and 4,083,579 relate to the treatment of porous glass with non-radioactive ions (radioactive ions in the case of 3,364,148) followed by heating to close the pores which contain the ions. U.S. Pat. Nos. 3,147,225 discloses refractory particles, which contain no or minor amounts of silica and preferably are crystalline, within which particles a specifically selected radioactive cation is firmly fixed for use in self-luminous markers, liquid level indicators and other applications. There is no disclosure or suggestion in any of these patents, however, of reacting a silicate glass or silica gel having silicon-bonded alkali metal oxy, Group Ib metal cation oxy and/or ammonium cation oxy groups with radioactive wastes which typically are complex mixtures of radioactive and non-radioactive compounds and/or ions nor is there any disclosure of procedures for treating radioactive wastes.
U.S. Pat. No. 3,116,131 discloses the method of binding expanded silica particles with a binder and shaping and curing into a desired form, followed by impregnation with a solid desiccant, e.g., sodium hydroxide, and followed by impregnation with a radioactive gas and steam to absorb the water vapor followed by capillary condensation thereby entraining the radioactive gas in the pores after which the pores are closed by heating. The patent fails to disclose the chemical combination of waste radioactive ions in the pores of a porous glass or silica gel by ion exchange followed by closing the pores to provide the dual security of chemical combination and mechanical entrapment of the radioactive ions in the glass.
U.S. Pat. No. 3,959,172 discloses the method of forming and reacting a mixture of silicate or other source of silicon, a radionuclide waste and a metal cation to produce complex metalosilicate crystals which entrap the radionuclide waste. U.S. Pat. Nos. 3,451,940 and 3,849,330 disclose the utilization of a thermite reaction to form a complex polysilicate product containing the radioactive wastes. None of these patents teach or suggest the use of porous glass or silica gel which is characterized by an interconnecting porous structure of high ion-exchange capability such that upon contact with pernicious liquid radwaste the radioactive species (cations, for the most part) thereof become "ion-exchanged" to the glass or silica gel, and/or followed by treating to collapse the porous structure thereof to immobilize such radioactive species.
U.S. Pat. No. 3,167,504 discloses the purification of radioactive waste liquid by absorption on a synthetic zeolite which is then sealed in a suitable container for burial. There is no disclosure of the porous glass or silica gel discussed previously and/or of heating to collapse the pores of a porous glass or silica gel around radioactive materials.
U.S. Pat. Nos. 3,114,716; 3,262,885; 3,365,578 and 4,020,004 each deal with various techniques involving the preparation of glass-forming mixtures followed by firing to form a glass and none of them disclose the impregnation of a porous glass or silica gel with radioactive wastes whereby the radioactive species thereof become chemically bonded to the porous glass or silica gel followed by heating to collapse the pores.
U.S. Pat. No. 3,093,593 discloses methods for disposing of radioactive wastes by forming a porous ceramic piece from clays and other silicates followed by prefiring such pieces to destroy ion exchange capacity and thereafter impregnating the prefired pieces with radioactive liquid wastes. The pieces saturated with radioactive waste are then heated to vitrify them and render them non-absorptive. This patent teaches away from our invention which relies on ion exchange for strongly binding the radioactive waste in the pores of a glass or silica gel.
U.S. Pat. No. 3,938,974 relates to glass, optical wave guide fibers and their production. Radioactive materials cannot be used in such fibers because they form color centers which absorb light. Not only does this patent fail to disclose the use of radioactive materials, the presence of such materials are inimical to the express objects of the patent.
Two articles, The Behavior of Silica Gel Towards Certain Aklalies and Salts in Aqueous Solution by W. A. Patrick and E. H. Barclay, J. Phys. Chem., Volume 29, page 1400, 1925, and Interaction of Metal Silica Gels With Aqueous Electrolyte Solutions by L. V. Ponomareva, A. P. Dushina and V. B. Aleskovskii, Zhurnal Prikladnoi Khimii, Vol. 48, No. 10, pp. 2150-2155, 1975 relate to the ion exchange of alkali metal cations in a silica gel with heavy metal cations. However, neither of these articles are concerned with methods to remove radioactive species from radioactive waste streams by impregnating porous glass or silica gel with such streams, followed by heating to collapse the pores.
As will be apparent hereinafter from the various aspects of applicant's contributions to the art, there are provided novel methods to obtain novel compositions and articles for the containment of pernicious and dangerous radioactive materials over extraordinarily long periods of time. Unlike melting glass containment procedures, the methods of the invention need not involve any steps which would expose radioactive material to high temperatures, e.g., above about 900.degree. C., thereby eliminating the environmental hazard due to possible volatilization of radioactive material into the atmosphere.
Belgian Pat. No. 839,705, issued July 16, 1976 and German Offenlegungsschrift No. 2,611,495, published July 10, 1976 correspond substantially to U.S. Pat. No. 4,110,096, issued Aug. 29, 1978 to Pedro B. Macedo named as an inventor herein and Theodore A. Litovitz. These patents and Offenlegungsschrift contain essentially the same disclosures but there is no disclosure of porous glass forms having sufficient ion exchange capabilities to bind practical amounts of radioactive cations to the glass to thereby concentrate and contain said radioactive cations in the manner taught herein. Although mention is made of removing silica gel, that may have deposited in the pores of the porous glass during its manufacture, by washing with sodium hydroxide, it has been found that procedures for doing so known in the art fail to produce sufficient amounts of silicon-bonded sodium oxy (ion exchange) groups needed for providing the binding of practical levels of radioactive cations from the radwaste. Furthermore, the presence of silica gel in the pores can be advantageous in this invention as providing more surface area and a higher proportion of silicon-bonded hydroxyl groups and ultimately higher amounts of silicon-bonded metal oxy or ammonium oxy groups for ion exchange with radioactive cations.